[Federal Register Volume 59, Number 234 (Wednesday, December 7, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-29925]
[[Page Unknown]]
[Federal Register: December 7, 1994]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating Licenses
Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from November 14, 1994, through November 25,
1994. The last biweekly notice was published on November 23, 1994 (59
FR 60377).
Notice Of Consideration Of Issuance Of Amendments To Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, And Opportunity For A Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and should cite the publication date and page
number of this Federal Register notice. Written comments may also be
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.
The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By January 6, 1995, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington DC, by the above
date. Where petitions are filed during the last 10 days of the notice
period, it is requested that the petitioner promptly so inform the
Commission by a toll-free telephone call to Western Union at 1-(800)
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator
should be given Datagram Identification Number N1023 and the following
message addressed to (Project Director): petitioner's name and
telephone number, date petition was mailed, plant name, and publication
date and page number of this Federal Register notice. A copy of the
petition should also be sent to the Office of the General Counsel, U.S.
Nuclear Regulatory Commission, Washington, DC 20555, and to the
attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick
County, North Carolina
Date of amendments request: October 25, 1994
Description of amendments request: The proposed change would delete
the remainder of Appendix B, Environmental Technical Specifications,
including section 2/3.3.1, Water Level in the Discharge Canal, and
Section 2/3.4, Meteorology.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Hydraulic - Water level in the Discharge Canal
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated. The deletion of the discharge canal
specification of 4.5 [plus or minus] 1 ft mean sea level (msl) with
daily monitoring still leaves the operating restrictions delineated
in Updated Final Safety Analysis Report (UFSAR) Section 2.4.8.3.3
(4.5 [plus or minus] 2 ft msl) and the National Pollutant Discharge
Elimination System (NPDES) permit requirements to minimize the
impact of the discharge canal on the local groundwater supply. As
stated in this UFSAR section, the effect on the local ground water
regime will be minimal within this band. Level recorders in the
control room facilitate the continued monitoring the discharge canal
level in excess of the Appendix B Environment Technical
Specification (ETS) listed daily surveillance. This change in no way
affects the design or operation of equipment that could initiate or
mitigate any accident previously evaluated.
2. The proposed amendment would not create the possibility of a
new or different kind of accident from any accident previously
evaluated. The discharge canal level is an environmental concern
with the effects of a spill of radioactive liquids (UFSAR Section
2.4.12.3) being a path to the intake canal, due to the areas natural
gradient. With the estimated travel time for the liquid to reach the
canal (intake) at 60 years, and the large flow rate the degree of
dilution is such that this does not pose a threat to local wells.
The amendment would not affect the operation or design of any plant
equipment; therefore, no new credible accidents are created. In
addition, the proposed amendment would not affect the capability of
the response systems to mitigate the consequences of any accident
previously evaluated; therefore, no new or different accident would
result from this change.
3. The proposed amendment does not involve a significant
reduction in a margin of safety. The existence of these ETS does not
provide a margin of safety related to the nuclear operation of the
site. No safety limits are affected by this change. Therefore, this
amendment would not result in a reduction in any margin of safety.
Meteorology
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated. The deletion of the meteorology specification
still leaves the program delineated in UFSAR Section 2.3.3, Onsite
Meteorological Measurements Program, and reporting/recording of the
hourly meteorological data required to support Technical
Specification 6.9.1.10.a, Semiannual Radioactive Effluent Release
Report. This program is based on the meteorological monitoring
program described in Regulatory Guide 1.23, and NUREG-0654. While
the existing ETS 30 day reporting requirement for extended out-of-
service time and shiftly manual acquisition of data during batch or
accidental releases are not otherwise covered, the program does
contain the Regulatory Guide 1.23 reference to 90% data recovery and
a backup phone line is available for data retrieval. This proposed
amendment in no way affects the design or operation of equipment
that could initiate or mitigate any accident previously evaluated.
2. The proposed amendment would not create the possibility of a
new or different kind of accident from any accident previously
evaluated. The meteorological monitoring program is specified in
UFSAR Section 2.3.3. The amendment would not affect the operation or
design of any plant equipment; therefore, no new credible accidents
are created. In addition, the proposed amendment would not affect
the capability of the response systems to mitigate the consequences
of any accident previously evaluated nor would the amendment reduce
the effectiveness of the Emergency Response Plan; therefore, no new
or different accident would result from this change.
3. The proposed amendment does not involve a significant
reduction in a margin of safety. The meteorological ETS does not
provide any additional margin of safety related to the operation of
a nuclear plant. The meteorological program established in the UFSAR
covers the requirements stated in Appendix E to 10 CFR 50 by
providing meteorological systems adequate for determining the
magnitude of, and for continuously assessing the impact of, the
release of radioactive materials to the environment. The
meteorological instrumentation is used to measure environmental
parameters which may affect distribution of fission products and
gases following a Design Basis Accident (DBA); however, it is not a
primary success path for the mitigation of a DBA. No safety limits
are affected by this change. Therefore, this amendment would not
result in a reduction in any margin of safety.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
NRC Project Director: William H. Bateman
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick
County, North Carolina
Date of amendments request: October 28, 1994
Description of amendments request: The proposed changes would
revise the Technical Specifications to increase the surveillance test
intervals and allowable out-of-service times for selected
instrumentation addressed in Section 3/4.3. The proposed changes would
permit specified channel functional tests to be conducted quarterly
rather than weekly or monthly. Specifically, the proposed changes would
revise the surveillance test intervals and allowable out-of-service
times for the reactor protection system instrumentation, isolation
actuation instrumentation, emergency core cooling system actuation
instrumentation, control rod withdrawal block instumentation, control
room emergency ventilation system instrumentation, anticipated
transient without scram - recirculation pump trip system
instrumentation, end-of-cycle recirculation pump trip system
instrumentation, and reactor core isolation cooling system actuation
instrumentation, in accordance with NRC-approved General Electric
Company Licensing Topical Reports and NUREG-1433, Standard Technical
Specifications, General Electric Plants, BWR/4.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1) Operation of Brunswick Steam Electric Plant, Units 1 and 2,
in accordance with the proposed amendment, would not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The generic Licensing Topical Report, NEDC-30851P-A, assessed
the impact of changing RPS surveillance test intervals (STIs) and
allowable out-of-service times (AOTs) on the RPS failure frequency,
the scram frequency and equipment cycling. Specifically, Section
5.7.4, ``Significant Hazards Assessment'' of NEDC-30851P-1 states:
``Fewer challenges to the safeguards system, due to less
frequent testing of the RPS, conservatively results in a decrease of
approximately one percent in core damage frequency''. This decrease
is based upon the following:
* Based on the plant-specific experience presented in Appendix
J, the estimated reduction in scram frequency (0.3 scrams/yr)
represents a 1 to 2 percent decrease in core damage frequency based
on the BWR plant specific Probabilistic Risk Assessments (PRAs)
listed in Table 5-8.
* The increase in core damage frequency due to less frequent
testing is less than one percent. This increase is even lower (less
than 0.01 percent) when the changes resulting from the
implementation of the Anticipated Transients Without Scram (ATWS)
rule are considered. Therefore, this increase is more than offset by
the decrease in CDF due to fewer scrams.
* The effect of reducing unnecessary cycles on RPS equipment,
although not easily quantifiable also results in a decrease in core
damage frequency.
* The overall impact on core damage frequency of the changes in
allowable out-of-service time is negligible.''
From this generic analysis, the BWR Owners' Group concluded and
CP&L concurs that the proposed changes do not significantly increase
the probability or consequences of an accident previously evaluated,
since the increase in probability of a a scram failure due to RPS
unavailablity is insignificant. The overall probability of an
accident is decreased as the time the RPS instrumentation logic
operates undisturbed is increased, resulting in fewer inadvertent
scrams during testing and repair. The proprietary plant-specific
analysis contained in this submittal (Enclosure 6) demonstrates
that, although BSEP Units 1 and 2 differ from the generic plant
analyzed in LTR NEDC-30851P-A, the net effect of the plant-specific
differences does not alter the generic conclusions.
The generic Licensing Topical Reports, NEDC-30851P-A, Supplement
2 and NEDC 31677P-A, assessed the impact of changing STIs and AOTs
for BWR Isolation Instrumentation. Section 4.0, ``Summary of
Results,'' of NEDC-30851P-A, Supplement 2 states:
``The results indicate that the effects on probability of
failure to initiate isolation are very small and the effects on
probability or frequency of failure to isolate are negligible in
nearly every case. In addition, the results indicate that increasing
the AOT to 24 hours for tests and repairs has a negligible effect on
the probability of failure of the isolation function. These combined
with changes to the testing intervals and allowable out-of service
times for RPS and ECCS instrumentation provide a net improvement to
plant safety and operations.''
and Section 5.6, ``Assessment of Net Effect of Changes,'' of
NEDC-31677P-A states:
``A reduction in core damage frequency (CDF) of at least as much
as estimated in the ECCS instrumentation analysis can be expected
when the isolation actuation instrumentation STIs are changed from
one month to three months. The chief contributor to this reduction
is the channel functional tests for the MSIVs. Inadvertent closure
of the MSIVs will cause an unnecessary plant scram. This reduction
in CDF more than compensates for any small incremental increase (10%
or 1.0E-07/year) in calculated isolation function failure frequency
when the STI is extended to three months.''
From this generic analysis, the BWR Owners' Group concluded and
CP&L concurs that the proposed changes do not significantly increase
the consequences of an accident previously evaluated, since the
increase in probability of an isolation failure due to isolation
instrumentation unavailability is insignificant. For those
parameters common to RPS, the overall probability of an accident is
actually decreased as the time the RPS instrumentation logic
operates undisturbed is increased, resulting in less inadvertent
scrams during testing and repair. The plant-specific evaluation
provided with this submittal (Enclosure 8) demonstrates that the
conclusions of the generic analyses are applicable to BSEP Units 1
and 2.
The generic Licensing Topical Report, NEDC-30936P-A (Parts 1 and
2), assessed the impact of changing STIs and AOTs for all BWR ECCS
Actuation Instrumentation. Section 4.0, ``Technical Assessment of
Changes,'' of NEDC-30936P-A (Part 2) states:
``The results indicate an insignificant (less than 5E-7 per
year) increase in water injection function failure frequency when
STIs are increased from 31 days to 92 days, AOTs for repair of the
ECCS actuation instrumentation are increased from one hour to 24
hours, and AOTs for surveillance testing are increased from two to
six hours. For all four BWR models the increase represents less than
4% increase in failure frequency. However, when other factors which
influence the overall plant safety are considered, the net result is
judged to be an improvement in plant safety.''
From this generic analysis, the BWR Owners' Group concluded and
CP&L concurs that the proposed changes do not significantly increase
the probability or consequences of an accident previously evaluated,
since the increase in probability of a water injection failure due
to ECCS instrumentation unavailability is insignificant and the net
result is judged to be an improvement in plant safety. The plant-
specific analysis contained in this submittal (Enclosure 7)
demonstrates that, although BSEP Units 1 and 2 differ from the
generic model analyzed in LTR NEDC-30936P-A, the net effect of the
plant-specific differences does not alter the generic conclusions.
The generic Licensing Topical Reports, NEDC-30851P-A, Supplement 1,
and GENE-770-06-1-A assessed the impact of changing Control Rod
Block STIs and AOTs on Rod Block failure frequency. GENE-770-06-1-A
also assessed the impact of changing STIs and AOTs on ATWS-RPT and
EOC-RPT failure frequency. Section 5 (Brookhaven National
Laboratory's Technical Evaluation Report - Attachment 2 to the NRC
SER) of NEDC-30851P-A, Supplement 1 states:
1``The BWR Owners' Group proposed changes to the Technical
Specifications concerning the test requirements for BWR control rod
block instrumentation. The changes consist of increasing the
surveillance test intervals from one to three months. These test
interval extensions are consistent with the already approved changes
to STIs for the Reactor Protection System. The technical analysis
reviewed and verified as documented herein indicates that there will
be no significant changes in the availability of the control rod
block function if these changes are implemented. In addition, there
will be a negligible impact on the plant core melt frequency due to
the decreased testing.''
and Section 2.0, ``Summary'' of GENE-770-06-1-A states:
``Technical bases are provided for selected proposed changes to
the instrumentation STIs and AOTs that were identified in the BWROG
Improved BWR Technical Specification activity. These STI and AOT
changes are consistent with approved changes to the RPS, ECCS, and
isolation actuation instrumentation. These proposed changes do not
result in a degradation to overall plant safety.''
Based on the generic analysis in NEDC-30851P-A, Supplement 1,
the BWR Owners' Group concluded and CP&L concurs that the proposed
changes to Control Rod Withdrawal Block instrumentation do not
significantly increase the probability or consequences of an
accident previously evaluated. Also, based on the generic assessment
in GENE-770-06-1-A, the BWR Owners' Group concluded and CP&L concurs
that the proposed changes to the ATWS-RPT and EOC-RPT
instrumentation do not significantly increase the probability or
consequences of an accident previously evaluated.
Bases contained in GE Topical Report GENE-770-06-2P-A, assessed
the impact of changing STIs and AOTs on BWR RCIC failure frequency.
Section 2.0, ``Summary'' of GENE-770-06-2P-A states:
``The STI and AOT changes to the RCIC actuation instrumentation
are justified based on their small effect on the water injection
function unavailability and consistency with comparable changes to
actuation instrumentation for the other ECCS subsystems.''
On this basis, the BWR Owners' Group concluded and CP&L concurs
that the proposed changes to RCIC instrumentation do not
significantly increase the probability or consequences of an
accident previously evaluated.
2) Operation of Brunswick Steam Electric Plant, Units 1 and 2,
in accordance with the proposed amendment, would not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed changes do not alter the physical characteristics
or function of any plant systems or components and they do not
introduce any new mode of operation. Therefore, system and component
performance would not be challenged in a manner that could create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3) Operation of Brunswick Steam Electric Plant, Units 1 and 2,
in accordance with the proposed amendment, would not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed and approved the generic studies
contained in the LTRs and has concurred with the BWR Owners' Group
that the proposed changes do not significantly affect the
probability of failure or availability of the affected Instrument
Systems. The proposed changes to AOTs provide realistic times to
complete the required actions without increasing the overall
instrument failure frequency. Likewise, the extended STIs do not
result in significant changes in the probability of instrument
failure. Furthermore, the proposed changes will reduce the
probability of test-induced plant transients and equipment failures.
Finally, instrument setpoint drift will remain within present
tolerances, thereby assuring that the margin of safety, as
demonstrated by applicable safety analyses, remains unchanged.
Therefore, it is concluded that the proposed changes would not
result in a reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
NRC Project Director: William H. Bateman
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: November 4, 1994
Description of amendment request: The requested amendment will
change the testing frequency of the turbine overspeed protection valves
from monthly to quarterly.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The requested change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The requested change will have no influence on the
probability or consequences of an accident previously evaluated. The
accident of concern to this requested change is a turbine overspeed
with missile generation impacting safety related components or
structures. The evaluation in WCAP-11525 shows that the probability
of a missile ejection incident will not be affected with the
requested frequency reduction to the turbine overspeed protection
valve surveillance test. There is no change to the consequences of
the event as the postulated accident event is unchanged.
Accordingly, the requested change will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The requested change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated. The change affects the test interval for the turbine
overspeed protection valves and does not change the design,
operation, or failure modes of the valves and other components in
the turbine overspeed protection system. Therefore, the requested
change will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. The requested change does not involve a significant reduction
in the margin of safety. The probability of turbine overspeed with
an extension of the testing interval has been determined to be
within applicable acceptance criteria. The change does not affect
the design, operation, or failure modes of the valves or other
components in the turbine overspeed protection system. Accordingly,
the requested change will not involve a significant reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
NRC Project Director: William H. Bateman
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,
Will County, Illinois
Date of amendment request: November 7, 1994
Description of amendment request: The proposed amendment would
allow an increase to the allowable nominal fuel enrichment from 4.2 to
5.0 weight percent Uranium-235 (w/o U-235). The changes include: (1)
increasing the allowable storage enrichment in Region 1 and allowing
the use of Integral Fuel Burnable Absorbers (IFBAs) for reactivity
equivalencing, (2) revising the Region 2 discharge burnup curve to
include nominal fuel enrichments up to 5.0 w/o U-235, and (3) making
editorial changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes to Section 5 of Technical Specifications do
not affect any accident initiators or precursors and do not change
or alter the design assumptions for the systems or components used
to mitigate the consequences of an accident. The fuel enrichment
increase will not affect reactor operation or the core design
methods. The physical characteristics of the fuel assemblies are not
changed, and fuel assembly movement will continue to be controlled
by approved fuel handling procedures. Reload core designs will
continue to be performed on a cycle by cycle bases as part of the
reload safety evaluation process, using NRC approved codes and
methods. Each reload design is evaluated to confirm that the cycle
core design adheres to the limits that exist in the accident
analyses and Technical Specifications to ensure that reactor
operation is acceptable.
The proposed changes are consistent with the analysis performed
in the ``Criticality Analysis of Byron and Braidwood Station Fuel
Storage Racks.'' The analysis was revised in June 1994 to include
boraflex gaps and shrinkage. The revised analysis is provided in the
proposed Technical Specification amendment. The analysis methodology
has been previously accepted by the NRC and is consistent with the
appropriate standards to establish the Keff limit for storage
racks and to calculate the maximum Keff. The reanalysis
addresses the most limiting postulated accident of a misloaded fuel
assembly and has shown that having at least 300 ppm of soluble boron
offsets any positive reactivity impacts for any of the postulated
accidents. The concentration of boron in the spent fuel pool water,
which is administratively controlled, is sufficient to maintain
Keff less than or equal to 0.95. The analysis is bounding for a
dropped fuel assembly on top of a rack or between rack modules, loss
of cooling systems, and reduction the fuel pool temperature to less
than 50 deg.F. The proposed changes do not impact any other accident
previously evaluated in the [Updated Final Safety Analysis Report]
UFSAR. There is no postulated accident that could cause reactivity
to increase beyond the analyzed conditions in the spent fuel racks.
There is no impact on the ability of the Spent Fuel Pool cooling
system to maintain the bulk pool temperature within limits. The
UFSAR analysis performed to calculate the maximum fuel cladding
temperature and spent fuel pool cooling include assumptions which
bound the use of more highly enriched fuel assemblies. Although fuel
enrichment is not a specific assumption in any of these analyses,
the heat load of a typical core offload may change with higher
enrichments. The average burnup of the offload will be increased
since few assemblies will be used per cycle; however, the new heat
load will continue to be [bound] by the UFSAR analysis because the
spent fuel pool racks have been analyzed for a total core offload
with all fuel assemblies having 4.5 years of operating time.
The radiological consequences analysis continues to bound the
licensed fuel burnup and enrichment at Byron and Braidwood stations.
The radiological consequences analysis results are a function of the
core inventory of radioactive isotopes. Since the maximum fuel
burnup limits and fuel peaking factors will not be exceeded, the
assumed fission product inventory will remain valid; therefore, the
limits of 10 CFR [Part] 100 continue to be met. Additionally, Byron
and Braidwood addressed the issue of the impact on the radiation
levels at the pool surface to the worker during non-accident
conditions. These conditions are not changed as [a] result of this
submittal, because the average fuel assembly burnup limit (isotopic
inventory) and maximum power produced in each fuel assembly will not
be changed by the increased fuel enrichment.
B. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not affect the design or operation of
any system, structure, or component in the plant. There are no
changes to parameters governing plant operation; no new or different
type of equipment will be installed. Each reactor core design will
continue to meet all design requirements; operation of the core will
not be affected. No modifications to the spent fuel pool are being
pursued and the fuel parameters used in the analysis remain
bounding. The method and manner in which the fuel will be stored in
the spent fuel pool has not changed. The proposed changes ensure
that 17X17 (Optimized Fuel Assembly, VANTAGE 5, VANTAGE +, and
PERFORMANCE +) fuel assemblies can be safely stored, maintaining a
Keff less than or equal to 0.95 under full water density
conditions, in both Regions 1 and 2 of the spent fuel pool. All
design criteria and criticality acceptance criteria continue to be
met. The reanalysis addresses the most limiting postulated accident
(misloaded fuel assembly) and has shown that having at least 300 ppm
of soluble boron offsets any positive reactivity impacts for any of
the postulated accidents. The level of boron in the spent fuel pool
water, which is administratively controlled, is sufficient to
maintain Keff less than or equal to 0.95. The reanalysis to
increase the storage enrichment of fuel in Regions 1 and 2 of the
spent fuel pool does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Additionally, approval of this amendment will not create a new
accident with regards to the new fuel storage vault which is
designed to handle the increased enrichment. The Byron and Braidwood
new fuel vaults were previously analyzed using NRC accepted
criticality analysis methodology in June 1989. This analysis was
performed to increase the storage enrichment of the New Fuel Vault
to 5.0 w/o U-235. The New Fuel Vault analysis was submitted to the
NRC and is the current licensing basis.
C. The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed changes do not affect the margin of safety for any
Technical Specification. All reactor design criteria will continue
to be met. The methodologies used in the accident analyses have been
accepted previously by the NRC and all criticality acceptance
criteria have been met under all assumed conditions (normal and
accident). The design basis for preventing criticality outside the
reactor is that, including uncertainties, there is a 95 percent
probability at a 95 percent confidence level that the effective
neutron multiplication factor, Keff, of the fuel assembly array
will be less than 0.95 as recommended by ANSI 57.2-1983 and OT
Position Paper for Review and Acceptance of Spent Fuel Storage and
Handling Applications, dated April 14, 1978. The analyses for both
Regions 1 and 2 fuel storage were verified to meet the above design
basis.
The criticality analysis for Regions 1 and 2 has been revised to
allow for storage of fuel assemblies with enrichments up to 5.0 w/o
U-235. The proposed Technical Specification changes include those
changes necessary to maintain Keff less than or equal to 0.95,
including conservative allowances for uncertainties and biases, when
the pool is flooded with unborated water. The proposed changes
include a requirement for fuel assemblies with enrichments above 4.2
w/o U-235 to contain sufficient integral fuel burnable absorbers
such that the maximum reference fuel K infinity is less than or
equal to 1.470 in unborated water at 68 deg.F due to restrictions on
spent fuel storage. Should a postulated accident occur which causes
a reactivity increase in the Byron and Braidwood Spent Fuel Pools,
Keff will be maintained less than or equal to 0.95 due to the
presence of at least 300 ppm of soluble boron in the spent fuel
pool. The proposed changes do not affect any plant safety parameters
or setpoints.
The proposed changes ensure that the design basis for preventing
criticality in the fuel storage areas is preserved, and fuel cycle
designs will continue to be analyzed using NRC accepted codes and
methods to ensure the design bases are satisfied.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: For Byron, the Byron Public
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for
Braidwood, the Wilmington Township Public Library, 201 S. Kankakee
Street, Wilmington, Illinois 60481
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690
NRC Project Director: Robert A. Capra
Connecticut Yankee Atomic Power Company (CYAPCO), and Northeast
Nuclear Energy Company (NNECO), Docket Nos. 50-213 and 50-245,
Haddam Neck Plant, and Millstone Nuclear Power Station, Unit 1,
Middlesex County, and New London County, Connecticut
Date of amendment request: October 31, 1994
Description of amendment request: The proposed amendments would
renew the existing license conditions for both plants to implement and
maintain Integrated Implementation Schedule (IIS) Program Plans (the
Program Plan). The Program Plans provide a methodology to be followed
for scheduling plant modifications and engineering evaluations.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
YAPCO and NNECO have reviewed the proposed changes in accordance
with 10 CFR 50.92 and conclude that the changes do not involve a SHC
[significant hazards consideration]. The basis for this conclusion
is that the three criteria of 10 CFR 50.92(c) are not compromised.
The proposed changes do not involve an SHC because the changes would
not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
Operation of the facilities in accordance with these proposed
changes would require the implementation of the IIS methodology
described in the Program Plans. As such, it requires that CYAPCO and
NNECO establish an administrative means for tracking, prioritizing,
and scheduling NRC-required plant modifications and engineering
evaluations, and licensee identified plant improvement projects.
This methodology is intended to enhance plant safety by more
effectively controlling the number and scheduling of plant
modifications, thereby assuring that issues required for safe
operation of the plants receive priority and are completed in a
timely manner. Because the license conditions address only an
administrative scheduling mechanism, it does not affect directly the
design or operation of the plant. Therefore, no accident analyses
are affected and the proposed changes do not increase the
probability or consequences of any previously evaluated accident.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed license conditions establish a requirement related
to scheduling of modifications and engineering evaluations. Because
the license conditions address only an administrative scheduling
mechanism, they do not affect directly the design or operation of
the plants. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from those
previously evaluated.
3. Involve a significant reduction in the margin of safety.
The proposed license conditions renew administrative
requirements intended to enhance public safety and reliable plant
operation. The proposed license conditions do not affect any
accident analyses, directly modify the plant configurations, or
change the way the plants are operated. The methodologies are
intended to enhance plant safety by more effectively controlling the
number and scheduling of plant modifications, thereby assuring that
issues required for safe operation of the plants receive priority
and are completed in a timely manner. Because the license conditions
address only an administrative scheduling mechanism, they do not
affect directly the design or operation of the plants. Therefore,
the proposed changes do not involve a reduction in any margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Russell Library, 123 Broad
Street, Middletown, Connecticut 06457, for the Haddam Neck Plant, and
the Learning Resource Center, Three Rivers Community-Technical College,
Thames Valley Campus, 574 New London Turnpike, Norwich, CT 06360, for
Millstone Unit 1.
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270.
NRC Project Director: Phillip F. McKee
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: August 5, 1994, as supplemented on
November 17, 1994.
Description of amendment request: This amendment is an additional
followup to the amendment request of May 29, 1992, published in the
Federal Register on July 8, 1992 (57 FR 30242), which changed the
Technical Specifications (TSs) Section 1.0, Definitions, to accommodate
a 24-month fuel cycle and which proposed the extension of the test
intervals for specific surveillance tests. This amendment proposes
extending the surveillance intervals to 24 months for the following
additional surveillance tests:
(1) Charging Flow Instrumentation
(2) Containment Sump, Recirculation Sump, and Reactor Cavity
Continuous Level Instrument Channels
(3) Auxiliary Feedwater Flow Rate Channel
(4) Control Room Air Filtration System
(5) Post Accident Containment Venting System
(6) Liquid Rad-Waste Flow Channel
(7) Steam Generator Blowdown Flow Channel
(8) Liquid Waste Distillate Tank Level Channels
(9) Primary Water Storage Tank Level Instrumentation
(10) Flow Rate Monitors; Plant Vent (Unit 2) and Stack Vent (Unit
1)
(11) Stack Vent Noble Gas Activity Monitor (R-60)
(12) High Pressure Water Fire Protection System
(13) Fire Protection System Diesel Engine
(14) Electrical Tunnel, Diesel Generator Building, and Containment
Fan Cooler Fire Protection Spray Systems; (A) System Functional Test
and (B) Spray Header Visual Inspection
(15) Penetration Fire Barriers
(16) Smoke Detectors/Electrical Penetration Area Inside Containment
(17) Functional Testing of Containment Sump Pumps
The changes requested by the licensee are in accordance with
Generic Letter 91-04, ``Changes in Technical Specification Surveillance
Intervals to Accommodate a 24-Month Fuel Cycle.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Charging Flow Instrumentation
The proposed change does not involve a significant hazards
consideration since:
1. A significant increase in the probability or consequences of
an accident previously evaluated will not occur.
It is proposed that the channel calibration frequency for the
Charging Flow instrumentation be changed from 18 months (+25%) to
every 24 months (+25%).
A statistical analysis of channel uncertainty for a 30 month
operating cycle has been performed. Based upon this analysis it has
been concluded that sufficient margin exists to accommodate the
channel statistical error resulting from a 30 month operating cycle.
The existing margin provides assurance that plant protective actions
will occur as required. It is therefore concluded that changing the
surveillance interval from 18 months (+25%) to 24 months (+25%) will
not result in a significant increase in the probability or
consequences of an accident previously evaluated.
2. The possibility of a new or different kind of accident from
any accident previously evaluated has not been created.
The proposed change in operating cycle length due to an
increased surveillance interval will not result in a channel
statistical allowance which exceeds the current margin. Plant
equipment will provide protective functions to assure that Safety
Analysis limits are not exceeded. This will prevent the possibility
of a new or different kind of accident from any previously evaluated
from occurring.
3. A significant reduction in a margin of safety is not
involved.
The above change in surveillance interval resulting from an
increased operating cycle will not result in a channel statistical
allowance which exceeds current margin. This margin, which is
equivalent to the existing margin, is necessary to assure that
protective safety functions will occur so that Safety Analysis
limits are not exceeded.(2) Containment Sump, Recirculation Sump,
and Reactor Cavity Continuous Level Instrument Channels
The proposed change does not involve a significant hazards
consideration since:
1. There is no significant increase in the probability or
consequences of an accident.
It is proposed that the calibration and test frequency for the
Containment Sump, Recirculation Sump and Reactor Cavity continuous
level monitoring instrument channels be revised from every 18 months
(+25%) to 24 months (+25%).
A statistical analysis of channel uncertainty for a 30 month
operating cycle has been performed. Based upon this analysis it has
been concluded that sufficient margin exists to accommodate the
channel statistical error resulting from a 30 month operating cycle.
The existing margin provides assurance that plant protective actions
will occur as required. It is therefore concluded that changing the
surveillance interval from 18 months (+25%) to 24 months (+25%) will
not result in a significant increase in the probability or
consequences of an accident previously evaluated.
2. The possibility of a new or different kind of accident from
any previously analyzed has not been created.
The proposed change in operating cycle length due to an
increased surveillance interval will not result in a channel
statistical allowance which exceeds current margin. Plant equipment
will provide protective functions to assure that Safety Analysis
limits are not exceeded. This will prevent the possibility of a new
or different kind of accident from any previously evaluated from
occurring.
3. There has been no reduction in the margin of safety.
The above change in surveillance interval resulting from an
increased operating cycle will not result in a channel statistical
allowance which exceeds current margin. This margin is necessary to
assure that protective safety functions will occur so that safety
analysis limits are not exceeded.
(3) Auxiliary Feedwater Flow Rate Channel
The proposed change does not involve a significant hazards
consideration since:
1. There is no significant increase in the probability or
consequences of an accident.
It is proposed that the calibration frequency for the Auxiliary
Feedwater Flow Rate channel be revised from 18 months (+25%) to 24
months (+25%).
A statistical analysis of channel uncertainty for a 30 month
operating cycle has been performed. Based upon this analysis it has
been concluded that sufficient margin exists between the existing
technical specification limits and the licensing basis Safety
Analysis limit to accommodate the channel statistical error
resulting from a 30 month operating cycle. The existing margin
between the Technical Specification limit and the Safety Analysis
limit provides assurance that plant protective actions will occur as
required. It is therefore concluded that changing the surveillance
interval from 18 months (+25%) to 24 months (+25%) will not result
in a significant increase in the probability or consequences of an
accident previously evaluated.
2. The possibility of a new or different kind of accident from
any previously analyzed has not been created.
The proposed change in operating cycle length due to an
increased surveillance interval will not result in a channel
statistical allowance which exceeds the current margin between the
existing Technical Specification limit and the Safety Analysis
limit. Plant equipment, which will be set at (or more conservatively
than) Technical Specification limits, will provide protective
functions to assure that safety analysis limits are not exceeded.
This will prevent the possibility of a new or different kind of
accident from any previously evaluated from occurring.
3. There has been no reduction in the margin of safety.
The above change in surveillance interval resulting from an
increased operating cycle will not result in a channel statistical
allowance which exceeds the margin which exists between the current
Technical Specification limit and the licensing basis Safety
Analysis limit. This margin, which is equivalent to the existing
margin, is necessary to assure that protective safety functions will
occur so that Safety Analysis limits are not exceeded.
(4) Control Room Air Filtration System
The proposed change does not involve a significant hazards
consideration since:
1. There is no significant increase in the probability or
consequences of an accident.
It is proposed that the surveillance frequency for the Control
Room Air Filtration System be changed from every 18 months (+25%) to
every 24 months (+25%).
For the flow tests, data from 1986 to date indicates that the
Control Room Filtration System performed in an acceptable manner
when surveilled on an 18 month (+25%) basis. The only discrepancy
was due to a hardware error and was independent of the time between
surveillances. Per Generic Letter 91-04, this past test history
provides an adequate basis to conclude that an extended operating
cycle would have minimal impact upon the flow characteristics of the
Control Room Filtration System. The modification of the filtration
system in 1993 only enhanced system performance.
With regard to the absorbance properties of the charcoal,
previous test data highlights a problem occurring during the 1986-
1987 period which subsequent testing confirms was adequately
resolved.
With the 1993 modification which increased the carbon bed
thickness from 1'' to 4'', performance can only be enhanced.
Therefore, it is concluded that a significant increase in the
probability or consequences of an accident previously evaluated will
not be incurred by changing the surveillance interval from 18 months
(+25%) to 24 months (+25%).
2. The possibility of a new or different kind of accident from
any previously analyzed has not been created.
A review of past historical surveillance data over 7 years
indicates no failures which were time dependent. The modification,
which was performed in 1993, can only enhance performance of the
system. New fans, an increased charcoal bed thickness, and new HEPA
[high-efficiency particulate air] filters will increase the
reliability of the system. Thus, it is concluded that the
possibility of a new or different kind of accident than that
previously evaluated has not been created.
3. There has been no significant reduction in the margin of
safety.
Past test data validated the acceptability of the previous air
filtration system for an extended surveillance interval. The
modification performed in 1993 will only enhance the reliability and
performance of the air filtration system. Thus, it is concluded that
a significant reduction in the margin of safety is not involved.
(5) Post Accident Containment Venting System
The proposed change does not involve a significant hazards
consideration since:
1. There is no significant increase in the probability or
consequences of an accident.
It is proposed that the surveillance frequency for the Post
Accident Containment Venting system be revised from every 18 months
(+25%) to 24 months (+25%).
A review of past test history from 1986 to date indicates that
the Post Accident Containment Venting System performed in a
satisfactory manner when the surveillance period was 18 months
(+25%). There was one discrepant condition noted in the 1989 test,
which, based upon subsequent tests in 1991 and 1993, does not appear
to have been age related. The 1989 observation concerning a gasket
is considered to be a one time only event and unlikely to reoccur as
a result of extending the surveillance interval from 18 months
(+25%) to 24 (+25%).
An added consideration, in terms of safety significance, is the
fact that the Post Accident Containment Venting system is diverse
and redundant to the post accident hydrogen recombiners which are
themselves redundant and the primary means of reducing the post
accident hydrogen concentration within containment. The venting
system is not relied upon for containment pressure control.
Due to the satisfactory past test history of the venting system,
together with its secondary role as a means of controlling post
accident hydrogen concentration, it is concluded that a significant
increase in the probability or consequences of an accident
previously evaluated will not be incurred by changing the
surveillance interval from 18 months (+25%) to 24 months (+25%).
2. The possibility of a new or different kind of accident from
any previously analyzed has not been created.
A review of past historical surveillance data over 7 years
indicates no failures which are considered to be time dependent.
Although one discrepant condition was observed in the 1989 test it
was not repeated in subsequent surveillances. Per Generic Letter 91-
04, this constitutes a sufficient basis for revising the
surveillance interval from 18 months (+25) to 24 months (+25%). This
extension in the operating interval is not expected to have an
impact upon the availability of the system. Thus, it is concluded
that the possibility of a new or different kind of accident
previously evaluated has not been created.
3. There has been no reduction in the margin of safety.
As past test data validates the presumption that an extended
operating cycle will not impact the availability of the Post
Accident Containment Venting Systems, it is concluded that a
significant reduction in the margin of safety is not involved.
(6) Liquid Rad-Waste Flow Channel
The proposed change does not involve a significant hazards
consideration since:
1. There is no significant increase in the probability or
consequences of an accident.
It is proposed that the channel calibration frequency for the
Liquid Rad-Waste Flow Channel be revised from every 18 months
(+25%).
A statistical analysis of channel uncertainty for a 30 month
operating cycle has been performed. Based upon this analysis it has
been concluded that sufficient margin exists to accommodate the
channel statistical error resulting from a 30 month operating cycle.
The existing margin provides assurance that plant protective actions
will occur as required. It is therefore concluded that changing the
surveillance interval from 18 months (+25%) to 24 months (+25%) will
not result in a significant increase in the probability or
consequences of an accident previously evaluated.
2. The possibility of a new or different kind of accident from
any previously analyzed has not been created.
The proposed change in operating cycle length due to an
increased surveillance interval will not result in a channel
statistical allowance which exceeds current margin. Plant equipment
will be set to provide protective functions to assure that Safety
Analysis limits are not exceeded. This will prevent the possibility
of a new or different kind of accident from any previously evaluated
from occurring.
3. There has been no reduction in the margin of safety.
The above change in surveillance interval resulting from an
increased operating cycle will not result in a channel statistical
allowance which exceeds the allowable operating margin. This margin,
which is equivalent to the existing margin, is necessary to assure
that protective safety functions will occur so that Safety Analysis
limits are not exceeded.
(7) Steam Generator Blowdown Flow Channel
The proposed change does not involve a significant hazards
consideration since:
1. There is no significant increase in the probability or
consequences of an accident.
It is proposed that the channel calibration frequency for the
Steam Generator Blowdown Flow channel be revised from every 18
months (+25%) to every 24 months (+25%).
A statistical analysis of channel uncertainty for a 30 month
operating cycle has been performed. Based upon this analysis it has
been concluded that sufficient margin exists to accommodate the
channel statistical error resulting from a 30 month operating cycle.
The existing margin provides assurance that plant protective actions
will occur as required. It is therefore concluded that changing the
surveillance interval from 18 months (+25%) to 24 months (+25%) will
not result in a significant increase in the probability or
consequences of an accident previously evaluated.
2. The possibility of a new or different kind of accident from
any previously analyzed has not been created.
The proposed change in operating cycle length due to an
increased surveillance interval will not result in a channel
statistical allowance which exceeds the current margin. Plant
equipment will provide protective functions to assure that Safety
Analysis limits are not exceeded. This will prevent the possibility
of a new or different kind of accident from any previously evaluated
from occurring.
3. There has been no reduction in the margin of safety.
The above change in surveillance interval resulting from an
increased operating cycle will not result in a channel statistical
allowance which exceeds the allowable operating margin. This margin,
which is equivalent to the existing margin, is necessary to assure
that protective safety functions will occur so that Safety Analysis
limits are not exceeded.
(8) Liquid Waste Distillate Tank Level Channels
The proposed change does not involve a significant hazards
consideration since:
1. There is no significant increase in the probability or
consequences of an accident.
It is proposed that the channel calibration frequency for the
Liquid Waste Distillate Tank level of tanks 13 and 14 be revised
from every 18 months (+25%) to every 24 months (+25%).
A statistical analysis of channel uncertainty for a 30 month
operating cycle has been performed. Based upon this analysis it has
been concluded that sufficient margin exists to accommodate the
channel statistical error resulting from a 30 month operating cycle.
The existing margin provides assurance that plant protective actions
will occur as required. It is therefore concluded that changing the
surveillance interval from 18 months (+25%) to 24 months (+25%) will
not result in a significant increase in the probability or
consequences of an accident previously evaluated.
2. The possibility of a new or different kind of accident from
any previously analyzed has not been created.
The proposed change in operating cycle length due to an
increased surveillance interval will not result in a channel
statistical allowance which exceeds current margin. Plant equipment
will be set to provided protective functions to assure that Safety
Analysis limits are not exceeded. This will prevent the possibility
of a new or different kind of accident from any previously evaluated
from occurring.
3. There has been no reduction in the margin of safety.
The above change in surveillance interval resulting from an
increased operating cycle will not result in a channel statistical
allowance which exceeds the allowable operating margin. This margin,
which is equivalent to the existing margin, is necessary to assure
that protective safety functions will occur so that safety analysis
limits are not exceeded.
(9) Primary Water Storage Tank Level Instrumentation
The proposed change does not involve a significant hazards
consideration since:
1. There is no significant increase in the probability or
consequences of an accident.
It is proposed that the channel calibration frequency for the
Primary Water Storage Tank Level instrumentation be changed from
every 18 months (+25%) to 24 months (+25%).
A statistical analysis of channel uncertainty for a 30 month
surveillance has been performed. Based upon this analysis it has
been concluded that sufficient margin exists to accommodate the
channel statistical error resulting form a 30 month surveillance.
The existing margin provides assurance that plant protective actions
will occur as required. It is therefore concluded that changing the
surveillance interval from 18 months (+25%) to 24 months (+25%) will
not result in a significant increase in the probability or
consequences of an accident previously evaluated.
2. The possibility of a new or different kind of accident from
any previously analyzed has not been created.
The proposed change in surveillance interval will result in a
channel statistical allowance which can be accommodated over a 30
month operating cycle. Plant equipment, which will be set at (or
more conservatively than) Technical Specification limits, will
provide protective functions to assure that Safety Analysis limits
are not exceeded. This will prevent the possibility of a new or
different kind of accident from any previously evaluated from
occurring.
3. There has been no significant reduction in the margin of
safety.
The above changes in surveillance interval resulting from an
increased operating cycle will not result in a channel statistical
allowance which exceeds current margin. This margin is necessary to
assure that protective safety functions will occur so that Safety
Analysis limits are not exceeded.
(10) Flow Rate Monitors; Plant Vent (Unit 2) and Stack Vent
(Unit 1)
The proposed change does not involve a significant hazards
consideration since:
1. There is no significant increase in the probability or
consequences of an accident.
It is proposed that the calibration frequency for the flow rate
monitors for the Plant Vent (Unit 2) and the Stack Vent (Unit 1) be
revised from every 18 months (+25%) to every 24 months (+25%).
A statistical analysis of channel uncertainty for a 30 month
operating cycle has been performed. Based upon this analysis it has
been concluded that sufficient margin exists to accommodate the
statistical error resulting from a 30 month operating cycle. The
existing margin provides assurance that plant protective actions
will occur as required. It is therefore concluded that changing the
surveillance interval from 18 months (+25%) to 24 months (+25%) will
not result in a significant increase in the probability or
consequences of an accident previously evaluated.
2. The possibility of a new or different kind of accident from
any previously analyzed has not been created.
The proposed change in operating cycle length due to an
increased surveillance interval will not result in a statistical
allowance which exceeds the current margin. Plant equipment will be
calibrated to provide data to assure that safety analysis limits are
not exceeded. This will prevent the possibility of a new or
different kind of accident from an previously evaluated from
occurring.
3. There has been no reduction in the margin of safety.
The proposed change in the surveillance interval resulting from
an increased operating cycle will not result in a channel
statistical allowance which exceeds the allowable operating margin.
This margin, which is equivalent to the existing margin, is
necessary to assure that protective safety functions will occur so
that safety analysis limits are not exceeded.
(11) Stack Vent Noble Gas Activity Monitor (R-60)
The proposed change does not involve a significant hazards
consideration since:
1. There is no significant increase in the probability or
consequences of an accident.
It is proposed that the calibration frequency for the stack vent
noble gas activity monitor be revised from every 18 months (+25%) to
every 24 months (+25%).
The current monitor replaced the previous monitor and therefore
there is only one refueling cycle surveillance data available which
proved to be satisfactory. The vendor recommends a calibration
period based on user experience. Insofar as the 18 month (+25%)
surveillance has proven to be acceptable, extension to a 24 month
(+25%) cycle is consistent with the vendor's recommendation. Any
additional uncertainty generated due to the extended surveillance is
bounded by the uncertainty inherent in a grab sample taken once per
24 hours which is the required compensatory action should the
monitor be inoperable. Since setpoints for alarms are not critical
to either plant operation or safety, since extensive margin is
reflected between the setpoint and applicable limits, it is
concluded that any additional uncertainty involved in a longer
surveillance cycle will not result in a significant increase in the
probability or consequences of an accident previously evaluated.
2. The possibility of a new or different kind of accident from
any previously analyzed has not been created.
This monitor measures the activity of potentially radioactive
gaseous effluent through the stack vent. The alarm setpoints are set
at a point sufficiently above expected radioactivity levels to avoid
unnecessary alarms and, at the same time, far below discharge
limits. The purpose of the monitor is to annunciate in the event an
unexpected spike in radioactivity level should occur so that
corrective action can be taken prior to exceeding a discharge limit.
The margin that exists between the discharge limit and the setpoint
is more than sufficient to accommodate any drift that could be
practically expected in a 24 month (+25%) operating cycle.
In this capacity, the monitor does not have setpoints which are
critical to plant operation or safety. Readings are not used in a
quantitative manner nor is accuracy important. It is important that
the instrument remain operable and respond to step changes in
radioactivity level over the operating cycle. It is therefore
concluded that an extended operating cycle will not result in the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. There has been no reduction in the margin of safety.
Sufficient margin exists between plant setpoints and applicable
limits to accommodate any realistic drift projected to occur over a
30 month operating cycle. Furthermore, instrument indications are
not used in a quantitative manner nor is instrument accuracy of
importance. Therefore, it is concluded that no significant reduction
in the margin of safety will result from an extended operating
cycle.
(12) High Pressure Water Fire Protection System
The proposed change does not involve a significant hazards
consideration since:
1. There is no significant increase in the probability or
consequences of an accident.
It is proposed that the system functional test of the High-
pressure Water Fire Protection System be changed from every 18
months (+25%) to every 24 months (+25%).
This system is a static system which is not normally required to
operate. The main fire pumps are on standby and are not in operation
except for testing. Thus, almost no wear is induced as a function of
time except that which results from being in standby status which is
minimal and slow acting. Under these circumstances, extending the
operating cycle between surveillances would be expected to have
negligible affect upon system operability. It is therefore concluded
that there would be no significant increase in the probability or
consequences of an accident as a result of an extended interval
between surveillances.
2. The possibility of a new or different kind of accident from
any previously analyzed has not been created.
Extension of the plant operating cycle will primarily extend the
time the pumps are in standby capacity. The potential for system
deterioration is minimal under these circumstances. Any
deterioration that does occur will be slow acting with respect to
time. A significant deterioration would be detected by a monthly
pump operating test. Thus, an extended operating cycle is not
expected to create the possibility of a new or different kind of
accident form [from] any previously analyzed.
3. There has been no reduction in the margin of safety.
Extension of the operating cycle by several months only serves
to extend the period of time when the pumps are in standby status.
Any deterioration under these circumstances will be slow acting.
Significant deterioration would be detected by the monthly operating
test. Therefore, it is concluded that an extended interval between
surveillances will involve no significant reduction in the margin of
safety.
(13) Fire Protection System Diesel Engine
The proposed change does not involve a significant hazards
consideration since:
1. There is no significant increase in the probability or
consequences of an accident.
It is proposed that the Fire Protection System Diesel Engine
Functional test be changed from every 18 months (+25%) to every 24
months (+25%).
Except for periodic testing, the diesel is in a standby state
and not subject to operational stress. Periodic testing imposes
limited wear as evidenced by the absence of major repairs during
past maintenance. Extension of the operating cycle for several
months is expected to have virtually no impact upon diesel
operability. Monthly testing would detect any degradation. Thus it
is concluded that there would be no significant increase in the
probability or consequences of an accident as a result of an
extended interval between surveillances.
2. The possibility of a new or different kind of accident from
any previously analyzed has not been created.
Extension of the plant operating cycle will, for the most part,
only extend the time spent by the pumps in standby capacity. The
potential for system deterioration is minimal under these
circumstances. Any deterioration that does occur will be slow acting
with respect to time. Significant deterioration in performance would
be detected by the monthly pump operating test. Thus, an extended
operating cycle is not expected to create the possibility of a new
or different kind of accident from any previously analyzed.
3. There has been no reduction in the margin of safety.
Extension of the operating cycle by several months only serves
to extend the period of time when the pumps are in standby status.
Any deterioration under these circumstances will be slow acting and
significant deterioration would be detected by the monthly operating
test. Therefore, it is concluded that an extended interval between
surveillances will involve no significant reduction in the margin of
safety.
(14) Electrical Tunnel, Diesel Generator Building, and
Containment Fan Cooler Fire Protection Spray Systems; (A) System
Functional Test and (B) Spray Header Visual Inspection
The proposed change does not involve a significant hazards
consideration since:
1. There is no significant increase in the probability or
consequences of an accident.
(a) It is proposed that the functional test surveillance
interval for the Electrical Tunnel, Diesel Generator Building and
Containment Fan Cooler Fire Protection Spray Systems be changed from
every 18 months (+25%) to every 24 months (25%).
(b) It is proposed that the Spray Header visual inspection
interval be revised from every 18 months (25%) to 24 months (+25%).
Extension of the surveillance interval for Electrical Tunnel and
Diesel Generator Building Fire Protection System functional tests
will have virtually no impact upon the operability of these systems.
These systems are accessible during normal operation and other
sections of the Technical Specifications (4.14.A.1.g.(i) and
4.14.B.1.a(i)) require that the system valve tests be conducted on
an annual (12 month) basis. These annual tests would reveal any
system deterioration prior to the conclusion of the proposed
extended surveillance interval.
For the Fan Cooler Fire Protection System as well as the Spray
Header itself, evaluation of surveillance data from the past five
refueling outages indicates minor discrepancies which would not have
impaired system operability.
It is therefore concluded that extension of the proposed
surveillance interval will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The possibility of a new or different kind of accident from
any previously analyzed has not been created.
For the Electrical Tunnel and Diesel Generator Building,
extension of the surveillance interval will have a negligible affect
as other portions of the Technical Specifications require the same
surveillance on an annual basis. For the spray header and the fan
cooler fire protection system, historical surveillance data
validates operability over an 18 month (+25%) interval which lends
confidence to conclude that operability will be maintained over a 24
month (+25%) interval. It is therefore concluded that the
possibility of a new or different kind of accident from any accident
previously evaluated has not been introduced.
3. There has been no reduction in the margin of safety.
Extension of the surveillance for two systems will have minimal
impact as the Technical Specifications impose more frequent testing
for system valves on an annual basis. For the Spray Header and Fan
Cooler Fire Protection System, as well as the fire protection system
for the Diesel Generator Building and Electrical Tunnel, it can be
stated that these systems are static existing mainly in a standby
capacity under which little deterioration would be expected. Past
surveillance data validates system reliability. It is therefore
concluded that increasing the time interval between inspections
would not involve a significant reduction in the margin of safety.
(15) Penetration Fire Barriers
The proposed change does not involve a significant hazards
consideration since:
1. There is no significant increase in the probability or
consequences of an accident.
It is proposed that the visual inspection frequency of the
penetration fire barriers listed in the Technical Specifications be
changed from every 18 months (+25%) to every 24 months (+25%).
The fire barrier penetration seals are static devices existing
in standby status. Normal environmental conditions exist during
normal plant operations. The only deterioration expected would be
that due to aging in a normal ambient which would be minimal to non-
existent. Evaluation of unacceptable seals detected during
surveillances indicates that initial seal installation was faulty
and aging was not the cause. Surveillances during four refueling
outages confirm this evaluation. Accordingly, it is not expected
that the proposed change in surveillance interval will involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The possibility of a new or different kind of accident from
any previously analyzed has not been created.
Past surveillances indicate that time is not a predominate
failure mechanism. In the few unacceptable seals detected, the
initial installation procedure has been identified as the cause of
the problems. Since the seals are static devices which exist in a
standby condition and experience normal ambient conditions during
normal operation, this would be the expected conclusion. In
addition, the fire barriers are just one means of fire protection.
Other means of fire protection exist such as fire alarms, sprinklers
and heat detectors which provide defense in depth. Thus, it is
concluded that the proposed change in the surveillance interval will
not create the possibility of a new or different kind of accident
from that previously evaluated.
3. There has been no reduction in the margin of safety.
Aging has not been identified as a principle contributor to seal
failures. In addition, there exists additional means of fire
protection which provides defense in depth. Therefore, the proposed
change in surveillance intervals is not expected to involve a
significant reduction in the margin of safety.
(16) Smoke Detectors/Electrical Penetration Area Inside
Containment
The proposed change does not involve a significant hazards
consideration since:
1. There is no significant increase in the probability or
consequences of an accident.
It is proposed that the surveillance interval for the smoke
detectors located in the electrical penetration area inside
containment be revised from every 18 months (+25%) to every 24
months (+25%).
Based on data taken from six surveillances from 1984 through and
including 1993, these devices have proven to be highly reliable. No
test failures were observed during this period. Based on the
guidance contained in Generic Letter 91-04, this demonstration of
reliable performance provides an adequate basis to conclude that the
proposed extension in the surveillance interval will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The possibility of a new or different kind of accident from
any previously analyzed has not been created.
Only 3 of the 5 detectors are required during normal operation.
Past surveillance data from six refueling outages indicate that it
is reasonable to expect all 5 detectors will remain operable over
the extended operating cycle which provides margin. It is therefore
concluded that the possibility of a new or different kind of
accident from any accident previously evaluated has not been
created.
3. There has been no reduction in the margin of safety.
The proven reliability of these devices indicates that a
significant reduction in the margin of safety would not be involved
in extending the operating cycle to 24 months (+25%).
(17) Functional Testing of Containment Sump Pumps
The proposed change does not involve a significant hazards
consideration since:
1. There is no significant increase in the probability or
consequences of an accident.
It is proposed that the functional test of the Containment Sump
Pump be changed from every 18 months (+25%) to every 24 months
(+25%).
No credit is taken within the FSAR for the Containment Sump
Pumps as a means of mitigating the consequences of an accident.
During normal operation the pumps serve as a means of quantifying
leakage inside Containment and therefore serve a safety function in
terms of accident prevention. However, in this capacity they are
only one of several systems which are capable of serving this
function and their failure would not result in a loss of this
capability.
In addition, evaluation of surveillance data back to 1986
indicates, with one exception, that the devices are very reliable.
In one instance, the pumps did not actuate or cause operation within
the setpoint tolerance but did operate as required. This was
determined not to be a time dependent event.
It is therefore concluded that extending the interval between
refueling surveillances will not result in a significant increase in
the probability or consequences of an accident.
2. The possibility of a new or different kind of accident from
any previously analyzed has not been created.
Past surveillances indicate that time is not a predominate
failure mechanism. Also, there exists a Technical Specification
requirement to perform almost the same surveillance on a monthly
basis in addition to every refueling outage. This monthly test
diminishes any potential risk in extending the operating cycle. It
is therefore concluded that the possibility of a new or different
kind of accident from any previously analyzed has not been created.
3. There has been no reduction in the margin of safety.
Past surveillance data indicates that pump operation is
reliable. In addition, there are alternate means of providing the
safety function fulfilled by these pumps. Also, a monthly test is
required which would detect any malfunction prior to the end of an
extended operating cycle. It is therefore concluded that extending
the operating cycle by several months will not result in a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,
New York, New York 10003.
NRC Project Director: Michael J. Case, Acting
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: November 2, 1994
Description of amendment request: The proposed amendments will
upgrade existing TS 3/4.6.2.1 and TS 3/4.6.2.3 by adapting the combined
specification for Containment Spray and Cooling Systems, contained in
the Standard Technical Specifications for Combustion Engineering
Plants, to the St. Lucie units. The changes account for plant-specific
differences and include all related requirements of NUREG-1432, Rev. O,
specification 3.6.6A. Accordingly, the proposal is consistent with the
Commission's Final Policy Statement on Technical Specifications
Improvements (58 FR 39132).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Pursuant to 10 CFR 50.92, a determination may be made that a
proposed license amendment involves no significant hazards
consideration if operation of the facility in accordance with the
proposed amendment would not: (1) involve a significant increase in
the probability or consequences of an accident previously evaluated;
or (2) create the possibility of a new or different kind of accident
from any accident previously evaluated; or (3) involve a significant
reduction in a margin of safety. Each standard is discussed as
follows:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendment will upgrade the existing Limiting
Conditions for Operation (LCOs) associated with the Containment
Cooling and Spray Systems to be consistent with NUREG-1432, Standard
Technical Specifications for Combustion Engineering Plants. The
Containment Cooling and Spray Systems are not initiators of
accidents previously evaluated, but are included as part of the
success paths associated with mitigating various accidents and
transients. The redundancy afforded by Containment Cooling and Spray
Systems in conjunction with the requirements of the proposed LCO
assures that the safety function of these systems can be
accomplished considering single failure criteria. Neither the design
nor the safety function of the Containment Cooling and Spray Systems
have been altered, and the proposed amendment does not change the
applicable plant safety analyses. Therefore, operation of the
facility in accordance with the proposed amendment will not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendment will not change the physical plant or the
modes of operation defined in the facility license. The changes are
administrative in nature in that they do not involve the addition of
new equipment or the modification of existing equipment, nor do they
otherwise alter the design of St. Lucie Unit 1 & 2 systems.
Therefore, operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The safety function of the Containment Cooling System is to
provide containment heat removal during normal operation and
accident conditions. The safety function of the Containment Spray
System is to provide containment heat and iodine removal during
accident conditions. The proposed amendment, in conjunction with the
redundancy afforded by the Containment Cooling and Spray system
design, assures that these safety functions can be accomplished
considering single-failure criteria. The bases for required actions
and the action completion times specified for inoperable Containment
Cooling and Spray trains are consistent with the corresponding
specifications in NUREG-1432. The safety analyses for applicable
accidents and transients remain unchanged from those previously
evaluated and reported in the Updated Final Safety Analysis Report.
Therefore, operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
Based on the above discussion and the supporting Evaluation of
Technical Specification changes, FPL has determined that the
proposed license amendment involves no significant hazards
consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Attorney for licensee: Harold F. Reis, Esquire, Newman and
Holtzinger, 1615 L Street, NW., Washington, DC 20036
NRC Project Director: Mohan Thadani, Acting
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas
Date of amendment request: November 7, 1994
Description of amendment request: The proposed amendment would
change the number of diesel generators (emergency power supply)
required to be operable during Mode 5 with the loops filled and Mode 6
with greater than or equal to 23 feet of water above the reactor vessel
flange. In addition, changes to certain system specifications that are
affected by the changes for the emergency power supply were also
proposed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of accidents previously
evaluated.
The equipment which is affected by the technical specification
changes proposed here are not precursors to any accident postulated
to occur in Modes 5 and 6. Therefore, the probability of an accident
is not increased. A design review has demonstrated the ability of
the required systems to perform their accident mitigation functions
for the postulated accidents during Mode 5 and 6 operation.
Therefore, it is concluded that an increase in the consequences of
the postulated accidents will not result from the proposed Technical
Specifications.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
The system design, function, and performance is not affected by
these specifications. No new equipment interactions are created.
Calculations and Failure Modes and Effects Analyses (FMEA) have been
conducted for selected mechanical systems and show there are no
failures which would cause situations where applicable accidents
would not be mitigated or which would cause new accidents. On this
basis, the proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The electrical power system specifications support the equipment
required to be operable, commensurate with the current level of
safety, including the equipment requiring a diesel backed power
source. The design review results demonstrate that operation in
Modes 5 and 6, in accordance with the proposed Technical
Specification changes, is acceptable from an accident mitigation
standpoint. The basic Modes 5 and 6 plant system functions are not
changed. On this basis, the proposed change does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas 77488
Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,
P.C., 1615 L Street, NW., Washington, DC 20036
NRC Project Director: William D. Beckner
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas
Date of amendment request: November 7, 1994
Description of amendment request: The proposed amendment would
permit both containment personnel airlock doors to be open while moving
fuel during refueling operations.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed change to Technical Specification 3.9.4,
Containment Building Penetrations, would allow the containment
personnel airlock to be open during fuel movement and core
alterations. The containment personnel airlock is closed during fuel
movement and core alterations to prevent the escape of radioactive
material in the event of a fuel handling accident. The containment
personnel airlock is not an initiator to any accident. Whether the
containment personnel airlock doors are open or closed during fuel
movement and core alterations has no affect on the probability of
any accident previously evaluated.
The proposed change does alter assumptions previously made in
evaluating the radiological consequences of the fuel handling
accident inside the reactor containment building. The proposed
change allows for the containment personnel airlock to be open
during refueling. The radiological consequences described in this
change are bounded by those given in the South Texas Project Safety
Evaluation Report and General Design Criteria 19. All doses for the
proposed change are less than the acceptance criteria, therefore,
there is no significant increase in the consequences of an accident
previously analyzed.
The proposed change will significantly reduce the dose to
workers in the containment in the event of a fueling handling
accident by accelerating the containment evacuation process. The
proposed change will also significantly decrease the wear on the
containment personnel airlock doors and, consequently, increase the
reliability of the containment personnel airlock doors in the event
of an accident.
Since the probability of a fuel handling accident is unaffected
by the airlock door positions, and the increased doses do not exceed
acceptance limits, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed change affects a previously evaluated accident,
e.g., a fuel handling accident inside containment. The existing
accident has been modified to account for the containment personnel
airlock doors being opened at the time of the accident. It does not
represent a significant change in the configuration or operation of
the plant and, therefore, does not create the possibility of a new
or different type of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
The margin of safety is reduced when the offsite and control
room doses exceed the acceptance criteria in the STP SER. As
previously discussed in the response to question 1, the offsite and
control room doses are below the acceptance criteria. Therefore,
this proposed change does not significantly reduce the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas 77488
Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,
P.C., 1615 L Street, NW., Washington, DC 20036
NRC Project Director: William D. Beckner
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas
Date of amendment request: November 8, 1994
Description of amendment request: The proposed amendment would
require only one of the two battery chargers associated with each Class
1E 125 VDC Channel I and Channel IV to be operable.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of accidents previously
evaluated.
A single charger is able to maintain the operability of Channel
I or Channel IV at the design loading with a single failure
condition. The proposed change does not alter equipment or
assumptions made in previously evaluated accidents. The consequences
of previously evaluated accidents are not increased. On this basis,
the proposed change does not involve a significant increase in the
probability or consequences of accidents previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed change involves only the operability requirement
for the second battery charger in Channel I and Channel IV. The
failure modes and operating modes would then be identical for all
four STPEGS Class 1E DC channels. Failure modes and effects analyses
already performed for DC Channels II and III would thus become
applicable to Channels I and IV also. The change proposed by this
Technical Specification revision is bounded by the failure modes and
effects analysis provided as Table 8.3-8 of the STPEGS UFSAR
[Updated Final Safety Analysis Report]. On this basis, the proposed
change does not create the possibility of a new or different kind of
accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The proposed change involves only the operability requirement
for the second battery charger in Channel I and Channel IV. The
number and capacity of DC channels required is not affected by the
proposed change. The electrical loads supported by these DC channels
are not changed and the duration of their function is not impacted.
On this basis, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas 77488
Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,
P.C., 1615 L Street, NW, Washington, DC 20036
NRC Project Director: William D. Beckner
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas
Date of amendment request: November 8, 1994
Description of amendment request: The proposed amendment would
permit the substitution of an extended range neutron flux monitor for
one of the source range neutron flux monitors during refueling
operations.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident as previously
evaluated.
During refueling operations, the Source Range channels are used
only for monitoring changes in core reactivity, and does not provide
inputs for automatically actuated equipment. The same function could
be performed by an Extended Range channel. The combination of the
present Channel Check and the proposed Channel Calibration are
sufficient to ensure that the detectors are capable of monitoring
core reactivity changes. By providing the intended redundant core
reactivity monitoring, neither the possibility or consequences of an
accident previously evaluated are increased.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
During refueling operations, the Source Range Monitors are used
simply as monitoring instrumentation. Extended Range Monitors are
capable of performing this function. The combination of the present
Channel Check and the proposed Channel Calibration are sufficient to
ensure that the detectors are capable of monitoring core reactivity
changes.
The proposed change would require revision of STP refueling
procedures. However, the physical movement of fuel assemblies is
within the scope of current refueling procedures. No new mechanism
for fuel misloading or damage or boron dilution would be created by
the change. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change provides core reactivity monitoring
comparable to that provided by the use of the Source Range channels.
The Extended Range channel is capable of detecting core reactivity
changes and provides the intended redundancy. The combination of the
present Channel Check and the proposed Channel Calibration are
sufficient to ensure that the detectors are capable of monitoring
core reactivity changes. No margin of safety is compromised by this
change.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas 77488
Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,
P.C., 1615 L Street, NW., Washington, DC 20036
NRC Project Director: William D. Beckner
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine
YankeeAtomic Power Station, Lincoln County, Maine
Date of amendment request: October 24, 1994
Description of amendment request: The proposed amendment would
modify Technical Specifications Table 4.1-3 surveillance requirements
for new emergency feedwater flow instrumentation. Specifically, the
currently installed analog feedwater flow transmitters would be
replaced by new, digital-type flow transmitters. The new digital flow
emergency feedwater flow transmitters are continuously self-checking
and have a recommended calibration interval of 9 years. The licensee
proposes to verify flow whenever the system operates and send one
transmitter back to the manufacturer for recalibration every refueling
outage.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The staff's analysis is
presented below:
1. The proposed change does not involve a significant increase in
the
probability or consequences of an accident previously evaluated.
Performance of Technical Specifications Table 4.1-3 (items 10 a and
b) ensures the emergency feedwater flow transmitters are operable when
required. The proposed change will continue to ensure operabililty and
therefore will not increase the probability or consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Emergency feedwater flow transmitter operability verification is
maintained, with no change to the system's configuration. Thus, there
is no unique operating condition that could adversely affect system
functional performance.
3. The proposed change does not involve a significant reduction in
a margin of safety.
There is no change to any Final Safety Analysis Report Chapter 14
(Safety Analysis) event. There is no change to the demonstration of
component operability; thus, the proposed change does not involve a
significant reduction in a margin of safety.
Based on this review, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
Local Public Document Room location: Wiscasset Public Library, High
Street, P.O. Box 367, Wiscasset, Maine 04578
Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic
Power Company, 329 Bath Road, Brunswick, Maine 04011
NRC Project Director: Walter R. Butler
North Atlantic Energy Service Corporation, Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: October 7, 1994
Description of amendment request: The proposed amendment would
remove from the Technical Specifications certain audit responsibilities
of the Nuclear Safety Audit Review Committee and certain review
responsibilities of the Station Operation Review Committee relating to
the Emergency Plan and Security Plan and their implementing procedures.
The proposed changes are consistent with the guidance of Generic Letter
93-07.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below.
A. The changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated (10 CFR
50.92(c)(1)) because the proposed changes do not affect the manner by
which the facility is operated and do not change any facility design
feature or equipment. Since there is no change to the facility or
operating procedures, there is no affect upon the probability or
consequences of any accident previously analyzed.
B. The changes do not create the possibility of a new or different
kindof accident from any accident previously evaluated (10 CFR
50.92(c)(2)) because they do not affect the manner by which the
facility is operated. The proposed changes merely affect audit and
review responsibilities and their deletion or relocation to other
controlled documents does not introduce new or different accident
scenarios.
C. The changes do not involve a significant reduction in a margin
of safety (10 CFR 50.92(c)(3)) because the proposed changes do not
affect the manner by which the facility is operated or involve
equipment or features which affect the operational characteristics of
the facility.
Based on this review, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
Local Public Document Room location: Exeter Public Library, 47
Front Street, Exeter, NH 03833.
Attorney for licensee: Thomas Dignan, Esquire, Ropes & Gray, One
International Place, Boston, MA 02110-2624.
NRC Project Director: Phillip F. McKee
Northeast Nuclear Energy Company (NNECO), Docket No. 50-245,
Millstone Nuclear Power Station, Unit 1, New London County,
Connecticut
Date of amendment request: October 14, 1994
Description of amendment request: The proposed change clarifies the
low pressure coolant injection (LPCI) requirements as required by
Technical Specification 4.5.A.2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
NNECO has reviewed the proposed change in accordance with 10 CFR
50.92 and concluded that the change does not involve a significant
hazards consideration (SHC). The basis for this conclusion is that
the three criteria of 10 CFR 50.92(c) are not compromised. The
proposed change does not involve an SHC because the change would
not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
The LPCI flow surveillance requirement to demonstrate that three
pumps can deliver 15,000 gpm does not relate to any previously
analyzed accident. There are no accident scenarios which rely upon
three pumps or 15,000 gpm. The existing scenarios are more limiting
in that they rely on, at the most, two LPCI pumps. The actual
testing of the pumps in accordance with the [inservice testing] IST
program and Technical Specification 4.13 will not change. The
testing of the pumps currently performed demonstrates that LPCI will
function to mitigate the postulated accidents. Therefore, the
elimination of the requirement to demonstrate that three pumps can
deliver 15,000 gpm will not involve an increase in the probability
or consequence of any previously evaluated accident.
The elimination of a requirement to test the LPCI header
instrumentation can not result in an increase to the probability or
consequence of an accident, since no such instrumentation exists, or
is required.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed change will remove a requirement to perform a
mathematical evaluation that provides no safety benefit. There is no
change in the test methodology currently performed. All four LPCI
pumps are tested. Deleting the requirement to verify that three LPCI
pumps can produce 15,000 gpm flow does not create the possibility of
a new or different type of accident.
Deleting the requirement to test the LPCI spray header
instrumentation can not create the possibility of a new or different
kind of accident, since there is no LPCI header instrumentation.
This change corrects an error which was introduced by an earlier
License Amendment.
3. Involve a significant reduction in the margin of safety.
This change to the LPCI testing requirements does not change any
of the actual testing, or individual component requirements which
exist for the LPCI system. The change to remove the three pump,
15,000 gpm flow requirement eliminates the need to calculate a value
which provides no relevant information in ascertaining the ability
of the LPCI system to perform its required safety function. The
existing testing ensures performance of the LPCI subsystem in
accordance with the accident analysis requirements. The intent of
the Technical Specification Surveillance Requirement remains
unchanged. Elimination of the requirement to test the LPCI header
instrumentation corrects an error introduced in an earlier License
Amendment. No LPCI header instrumentation exists, therefore, no
credit was taken for such instrumentation in determining the margin
of safety.
This change can not involve a significant reduction in the
margin of safety since there are no changes to the surveillance
requirements for any of the individual components of the LPCI
system.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, CT 06360.
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270.
NRC Project Director: Phillip F. McKee
Northern States Power Company, Docket Nos. 50-282 and 50-306,
Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue
County, Minnesota
Date of amendment requests: October 17, 1994, as supplemented
October 27, 1994
Description of amendment requests: The proposed amendments would
change the submittal frequency of the Radioactive Effluent Release
Report from semiannual to annual in accordance with 10 CFR 50.36a.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated
The proposed license amendments are requested to implement a
revision to 10 CFR 50.36a. The requested amendment[s] does not alter
any administrative controls over radioactive effluents, nor do they
affect any accident evaluations. Also, the requested amendments do
not involve any physical alterations to the plant with respect to
radioactive effluents. The proposed changes would only affect the
reporting requirements concerning routine data for radioactive
effluents.
Therefore, the probability or consequences of an accident
previously evaluated are not affected by any of the proposed
amendments.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
analyzed
The proposed license amendments are requested to implement a
revision to 10C FR 50.36a. The requested amendment[s] does not alter
any administrative controls over radioactive effluents, nor do they
involve any physical alterations to the plant with respect to
radioactive effluents. Also, the requested amendments do not change
the method by which any safety-related system performs its function.
The proposed changes would only affect the reporting requirements
concerning routine data for radioactive effluents.
Therefore, the possibility of a new or different kind of
accident from any accident previously evaluated would not be
created.
3. The proposed amendment will not involve a significant
reduction in the margin of safety
The proposed license amendments are requested to implement a
revision to 10 CFR 50.36a. The requested amendment[s] does not alter
any administrative controls over radioactive effluents, nor do they
involve any physical alterations to the plant with respect to
radioactive effluents. The proposed changes would only affect the
reporting requirements concerning routine data for radioactive
effluents. The operation of systems and equipment remains unchanged.
Therefore, a significant reduction in the margin of safety would
not be involved.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and
Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: John N. Hannon
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of amendment request: October 25, 1994
Description of amendment request: The amendment would add to the
Susquehanna Units 1 and 2 Technical Specifications, isolation signals
to Table 3.6.3-1 for the containment isolation valves on the sample
lines for the containment radiation monitoring (CRM) and wetwell sample
lines. This change is based on the licensee's design change for
installation of a new CRM and wetwell sample system.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The addition of the new CRM and Wetwell Sample System does not
affect any of the postulated initiating events identified in Chapter
6 and 15 of the FSAR, the Design Assessment Report, the current
Reload Analysis or the NRC Safety Evaluation Report (NUREG 0776).
The new CRM and Wetwell Sample System with separate containment
sample lines is isolated from the primary containment under accident
conditions. The power and control-power to the CRM from the Class 1E
Division I and Division II sources is through electrical isolation
schemes so that failure(s) in the CRM under accident conditions is
isolated from the Class 1E systems.
The addition of a new CRM and Wetwell Sample System with
separate sample lines and isolation valves does represent a change
in the probability of occurrence of a malfunction of equipment. The
addition of the auxiliary relay to the Division I and Division II
CAC System containment isolation logic does represent the source of
another potential malfunction in the logic due to the additional
relay in the circuit. However, the increase in probability due to
the additional relay is considered to be so small or insignificant
that the change is within the error bounds associated with the
original design calculations and does not constitute a significant
increase in probability of the overall system malfunction.
Thus, the addition of a new CRM and Wetwell Sample System does
not significantly increase the probability of occurrence or the
consequences of an accident or malfunction of equipment important to
safety, as previously evaluated in the SAR.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
Chapter 6 and 15 of the FSAR, the Design Assessment Report, the
current Reload Analysis and NUREG-0776 were reviewed to determine if
the proposed action had the potential of creating a postulated
initiating event which was not within the spectrum of events which
transient or anticipated operational occurrences and accident
conditions were analyzed. The review did not identify a postulated
initiating event which would create the possibility for an accident
of a different type.
A random single failure in the CRM A or CRM B does not create a
malfunction of a different type. A random single failure in the
existing containment isolation circuitry, the new isolation valve
control circuitry or the new valve position indication circuitry for
the new containment isolation valves does not create a malfunction
of a different type. The consequences of random single failure of
the CRM or the CRM and Wetwell containment isolation valve isolation
signal, control and indication circuitry is the same as the existing
consequences.
Thus, the addition of a new CRM and Wetwell Sample System does
not create a possibility for an accident or malfunction of a new or
different type.
3. Involve a significant reduction in a margin of safety.
The operability of the primary containment isolation valves for
the sample lines to the new CRMs and Wetwell Sample Rack is governed
by Technical Specification Section 3/4.6.3 entitled ``Containment
Systems, Primary Containment Isolation Valves'' with Table 3.6.3-1
establishing the maximum isolation time. The bases for operability
of the primary containment isolation valves is to ensure that the
containment atmosphere is isolated from the outside environment in
the event of a release of radioactive material to the containment
atmosphere or pressurization of the containment. This is consistent
with GDC 54 through 57 of 10 CFR 50, Appendix A. The bases for the
containment isolation within the time limits specified in Table
3.6.3-1 is for those isolation valves designed to close
automatically to ensure that the release of radioactive material to
the environment is consistent with the assumptions used in the
analyses for a LOCA. The new CRM and Wetwell Sample Rack sample line
isolation valves are solenoid valves which close immediately on an
accident signal. The proposed action does not affect the operability
requirements of Section 3/4.6.3. The margin of safety as defined in
the Technical Specification for the containment isolation valves is
not affected.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037
NRC Project Director: John F. Stolz
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: November 11, 1994
Brief description of amendments: The proposed amendment would
modify the technical specifications (TS) by deleting accelerated
testing and special reporting requirements for CPSES Units 1 and 2
emergency diesel generators. These changes are based on Generic Letter
94-01, ``Removal of Accelerated Testing and Special Reporting
Requirements for Emergency Diesel Generators,'' dated May 31, 1994.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Deletion of the requirement for special reporting of EDG
failures has no relation to probability or consequences of
accidents. Therefore, deletion of the requirement for special
reporting of EDG failures does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
There are no initiating events in accidents previously evaluated
that involve testing of EDGs. Therefore, deletion of accelerated
testing of EDGs does not involve a significant increase in the
probability of an accident previously evaluated.
A reduction in the number of test starts decreases EDG component
stress and wear and decreases unavailability time for maintenance
and pre and post run checks. The resulting change in EDG reliability
and availability is an improvement toward ensuring the EDGs are
capable of fulfilling their functional requirement to provide
electric power for safe shutdown of the plant during loss of offsite
power. Furthermore, implementation of the maintenance rule
provisions for performance monitoring and root cause analysis for
failures as a basis for establishing corrective actions establish an
alternate reliability basis that is at least equivalent to that
established by accelerated testing. Therefore, deletion of
accelerated testing of EDGs does not involve a significant increase
in the consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Deletion of the requirement for special reporting of EDG
failures introduces no new failure modes for the EDGs or other plant
systems and therefore has no relation to creation of accidents.
Therefore, deletion of the requirement for special reporting of EDG
failures does not create the possibility of a new or different kind
of accident from any accident previously evaluated.
The frequency at which EDG testing occurs does not affect the
potential failure modes of the EDGs, which have already been
assessed in the CPSES design. Therefore, deletion of accelerated
testing of EDGs does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Acceptance limits and failure values are not affected by the
requirement for special reporting of EDG failures. Therefore,
deletion of the requirement for special reporting of EDG failures
does not involve a significant reduction in a margin of safety.
The margin of safety impact associated with accelerated testing
relates to EDG reliability and availability. A reduction in the
number of test starts decreases EDG component stress and wear and
decreases unavailability time for maintenance and pre and post run
checks. The resulting change in EDG reliability and availability is
an improvement toward ensuring the EDGs are capable of fulfilling
their functional requirement to provide electric power for safe
shutdown of the plant during loss of offsite power. Furthermore,
implementation of the maintenance rule provisions for performance
monitoring and root cause analysis for failures as a basis for
establishing corrective actions establish an alternate reliability
basis that is at least equivalent to that established by accelerated
testing. Therefore, deletion of accelerated testing of EDGs does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, Texas 76019
Attorney for licensee: George L. Edgar, Esq., Newman and
Holtzinger, 1615 L Street, NW., Suite 1000, Washington, DC 20036
NRC Project Director: William D. Beckner
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: November 11, 1994
Brief description of amendments: The proposed amendment would
provide for cycle-specific allowances to account for increases in the
Heat Flux Hot Channel Factor between monthly surveillances.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed changes provide for the use of cycle-specific
allowances to account for F2Qc(z) increases
between surveillances. No hardware or setpoint changes are involved;
therefore, the changes have no impact on the probability of
occurrence of any accident previously analyzed.
The proposed changes ensure that F2Qc(z)
remains within its limit. Thus, the changes do not increase the
consequences of any accident previously analyzed.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed changes provide for the use of a cycle-specific
allowances to account for F2Qc(z) increases
between surveillances. The proposed changes do not involve any
hardware or setpoint changes. Therefore the changes do not create
the possibility of a new or different kind of accident from any
accident previously analyzed.
3. Does the proposed change involve a significant reduction in a
margin of safety?
The proposed changes do not affect the failure values of any
system or any event acceptance criteria. Higher cycle-specific
allowances ensure that remains below its limit between surveillances
and within the bounds considered in the safety analyses. Therefore,
the proposed changes do not involve a reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, Texas 76019
Attorney for licensee: George L. Edgar, Esq., Newman and
Holtzinger, 1615 L Street, NW., Suite 1000, Washington, DC 20036
NRC Project Director: William D. Beckner
Wisconsin Public Service Corporation, Docket No. 50-305,
Kewaunee Nuclear Power Plant, Kewaunee County, Wisconsin
Date of amendment request: November 8, 1994
Description of amendment request: The proposed amendment would
revise the Kewaunee Nuclear Power Plant (KNPP) Technical Specifications
to allow application of a voltage-based repair limit for the steam
generator (SG) tube support plate (TSP) intersections experiencing
outside diameter stress corrosion cracking (ODSCC).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
This proposed change was reviewed in accordance with the
provisions of 10 CFR 50.92 to show no significant hazards exist. The
proposed change will not:
1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Testing of model boiler specimens for free span tubing (no TSP
restraint) at room temperature conditions show burst pressures in
excess of 5,000 psig for indications of ODSCC with voltage
measurements as high as 19 volts. Burst testing performed on five
intersections pulled from the Kewaunee SGs with up to a 2 volt
indication showed measured tube burst in the range of 9,537 to 9,756
psig. Burst testing performed on pulled tubes from other plants with
up to 7.5 volt indications show burst pressures in excess of 6,300
psi at room temperatures. Correcting for the effects of temperature
on material properties and the minimum strength levels, tube burst
capability significantly exceeds the safety factor requirements of
RG 1.121.
Tube burst criteria are inherently satisfied during normal
operating conditions due to the presence of the TSPS. Test data
indicates that tube burst cannot occur within the TSP, even for
tubes with through wall EDM notches 0.75 inch long, when the notch
is adjacent to the TSP. Since tube burst is precluded during normal
operating conditions, the criterion that must be satisfied to
demonstrate adequate tube integrity is a safety margin of 1.43 times
MSLB pressure differential. From Figure 3-2 of EPRI report TR-
100407, the BOC structural limit for 7/8 inch diameter tubing is 9.6
volts. Applying an allowance of 20% for NDE uncertainty and 50% for
crack growth rate over an operating cycle results in a voltage
repair limit of 5.6 volts. The proposed repair limit of 2 volts is
very conservative when compared to the 5.6 volts taking into account
the low average growth rates experienced at Kewaunee and the high
tube burst pressures.
Relative to the expected leakage during accident condition
loadings, a plant specific calculation was performed to determine
the maximum primary-to-secondary leakage during a postulated MSLB
event. The evaluation considered both pre-accident and accident
initiated iodine spikes. The results of the evaluation show that the
accident spike yielded the limiting leak rate. This case was based
on a 30 rem thyroid dose at the site boundary and initial primary
and secondary coolant activity levels of 1.0 uCi/gm and 0.1 uCi/gm
dose equivalent iodine-131, respectively. A leak rate of 34.0 gpm
was determined to be the upper limit for allowable primary to
secondary leakage in the SG in the faulted loop. The SG in the
intact loop was assumed to leak at a rate of 0.1 gpm (150 gpd).
Application of the voltage-based repair limit will be
supplemented with a projected EOC MSLB leakage calculation and
conditional burst probability assessment. The methodology for
performing these calculations will be consistent with that discussed
in the draft GL until final guidance is published. Should the
projected MSLB leakage be exceeded indications will be repaired or
removed from service until the projected leakage is less than or
equal to 34.0 gpm.
Application of the voltage-based repair limit will not adversely
affect SG tube integrity. Therefore, the proposed amendment will not
increase the probability or consequences of an accident previously
evaluated.
2) Create the possibility of a new or different kind of accident
from any previously evaluated.
Implementation of the proposed voltage-based repair limit will
not reduce the overall safety or functional requirements of the SG
tube bundles. The tube burst criteria will be satisfied during
normal operating conditions by the presence of the TSPs. The RG
1.121 criteria that must be satisfied during accident loading
conditions is 1.43 times MSLB differential pressure. Conservatively,
the existing data base of burst testing shows that the tube burst
margins can be satisfied with bobbin coil signal amplitudes of about
8.82 volts or less regardless of the depth of tube wall penetration.
The proposed repair criteria will be supplemented with a reduced
operating leakage requirement of 150 gpd average through either SG
to preclude the potential for excessive leakage during operating
conditions. The 150 gpd restriction will provide for timely leakage
detection and plant shutdown in the event of the occurrence of an
unexpected single crack resulting in leakage that is associated with
the longest permissible crack length. The operating leakage limit is
based on leak-before break considerations, critical crack length and
predicted leakage.
The SG tube integrity will continue to be maintained through
inservice inspections and primary-to-secondary leakage monitoring.
Therefore, the proposed change will not create the possibility of a
new or different kind of accident.
3) Involve a significant reduction in the margin of safety.
Application of the voltage-based repair criteria has been
demonstrated to maintain tube integrity commensurate with the RG
1.121 criteria. RG 1.121 describes a method acceptable to the staff
for meeting GDCs 2, 14, 15, 31 and 32. This is accomplished by
determining the limiting degradation of SG tubing as established by
inservice inspection, beyond which tubes should be removed from
service. Upon implementation of the repair criteria, even under the
worst case conditions, the occurrence of ODSCC at the TSPs is not
expected to lead to a SG tube rupture event during normal or faulted
conditions. The most limiting event would be a potential increase in
leakage during a MSLB event. Excessive leakage during a MSLB is
precluded by verifying that the expected EOC crack distribution of
ODSCC indications at TSP locations would result in an acceptably low
primary-to-secondary leakage. Therefore, the radiological
consequences from tubes remaining in service is a small fraction of
the 10 CFR 100 limits.
The combined effects of a LOCA plus SSE on the SGs were assessed
as required by GDC 2. This issue was addressed for the Kewaunee SGs
through the application of leak-before-break (LBB) principles to the
primary loop piping. Based on the results of this analysis, it is
concluded that the LBB is applicable to the Kewaunee primary loops
and, thus, the probability of breaks in the primary loop piping is
sufficiently low that they need not be considered in the structural
design basis of the plant. Excluding breaks in the primary loops,
the LOCA loads from the large branch lines were also assessed and
found to be of insufficient magnitude to result in SG tube collapse.
Based on these analysis results, no tubes are expected to collapse
or deform to the degree that secondary-to-primary in-leakage would
be increased over currently expected levels. On this basis no tubes
need to be excluded from the voltage-based repair criteria for
reasons of deformation resulting from combined LOCA and SSE
loadings.
Addressing the RG 1.83 considerations, implementation of the
voltage-based repair criteria will include a 100% bobbin coil probe
inspection of all TSP intersections with known ODSCC down to the
lowest cold leg TSP identified. This will be supplemented by a
reduced operating leakage limit, enhanced eddy current data analysis
guidelines, MRPC inspection requirements and a projected EOC voltage
distribution. It is concluded that the proposed change will not
result in a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Wisconsin
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin
54301.
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P. O. Box 1497, Madison, Wisconsin 53701-1497.
NRC Project Director: Leif J. Norrholm
Previously Published Notices Of Consideration Of Issuance Of
Amendments To Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, And Opportunity For A Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of application for amendment: August 8, 1994
Brief description of amendment request: The proposed amendment
would modify Technical Specification (TS) 4.0.5.a. to delete the
requirement to obtain prior written relief from the NRC for inservice
inspection (ISI) and inservice testing (IST) of components conducted
pursuant to 10 CFR 50.55a. This change would provide relief from the
ASME Code requirement in the interim between the submittal of a relief
request and the NRC's issuance of a safety evaluation regarding the
relief request. The change would allow the plant to operate in
accordance with a proposed relief request while the NRC staff completed
its review of the relief request. The licensee has also proposed to
modify TS 4.0.5.b. to add a definition for biennial or every-2-year
inspection and testing activities. The definition of biennial or every
2 years will be at least once per 731 days.Date of individual notice in
Federal Register: November 14, 1994 (59 FR 56558)
Expiration date of individual notice: December 14, 1994
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Notice Of Issuance Of Amendments To Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson
SteamElectric Plant, Unit No. 2, Darlington County, South
CarolinaDate of application for amendment: June 29, 1994
Brief description of amendment: The amendment deletes the
requirement to perform alternate train testing to demonstrate that
other, similar, safety-related components are operable when components
are found, or made, inoperable in the safety injection, residual heat
removal, and containment spray systems. The surveillance requirements,
which the licensee refers to as accelerated testing requirements,
affect the safety injection (SI) pumps, residual heat removal (RHR)
pumps, containment spray (CS), SI, RHR and CS flow paths.
Date of issuance: November 21, 1994
Effective date: November 21, 1994
Amendment No. 153
Facility Operating License No. DPR-23. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: August 3, 1994 (59 FR
39581) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 21, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Hartsville Memorial Library,
147 West College, Hartsville, South Carolina 29550
The Cleveland Electric Illuminating Company, Centerior Service
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania
Power Company, Toledo Edison Company, Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of application for amendment: June 29, 1992, as supplemented
February 22, 1994
Brief description of amendment: The amendment revised TS Sections
3/4.3, ``Instrumentation,'' 3/4.4.2, ``Safety/Relief Valves,'' and
associated Bases to increase the surveillance test intervals and
allowable out-of-service times for specific safety-related
instrumentation.
Date of issuance: November 22, 1994
Effective date: November 22, 1994
Amendment No. 67
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 13, 1994 (59 FR
17605) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 22, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois;
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power
Station,Units 1 and 2, Rock Island County, IllinoisDate of
application for amendments: October 15, 1992
Brief description of amendments: The proposed amendments would
revise the Dresden and Quad Cities Technical Specification (TS) 3/4.4
to revise the sodium pentaborate solution concentrations for the
Standby Liquid Control System (SLCS) storage tanks based on net
positive suction head test results.
Date of issuance: November 16, 1994
Effective date: November 16, 1994
Amendment Nos.: 130, 124, 151, and 147
Facility Operating License Nos. DPR-19, DPR-25, DPR-29, and DPR-30.
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: December 9, 1992 (57 FR
58245) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 16, 1994. No
significant hazards consideration comments received: No
Local Public Document Room location: For Dresden, The Morris Public
Library, 604 Liberty Street, Morris, Illinois 60450; For Quad Cities,
The Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
PointNuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: October 29, 1993, as
supplemented on March 28, 1994, and November 8, 1994.
Brief description of amendment: The amendment revises the
surveillance intervals for the Volume Control Tank Level Instrument,
the Containment High Range Radiation Monitors, the Safety Injection
System Electrical Loading, the Safety Injection System, and the Reactor
Coolant System Subcooling Margin Monitors to accommodate a 24-month
fuel cycle. These revisions are being made in accordance wih the
guidance provided by Generic Letter 91-04, ``Changes in Technical
Specification Surveillance Intervals to Accommodate a 24-Month Fuel
Cycle.''
Date of issuance: November 16, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 178
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 20, 1994 (59 FR
37067) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 16, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: August 25, 1994
Brief description of amendments: The amendments revise the testing
interval for auxiliary feedwater (AFW) system pumps from monthly to
quarterly on a staggered test basis. The amendments are consistent with
the guidance in NUREG-1366, ``Improvements to Technical Specifications
Surveillance Requirements'' and Generic Letter 93-05, ``Line-Item
Technical Specifications Improvements to Reduce Surveillance
Requirements for Testing During Power Operation.'' In addition, a note
is incorporated from NUREG-1431, ``Revised Standard Technical
Specifications, Westinghouse Plants'' into the TS clarifying that the
turbine-driven AFW pump cannot be tested until the required pressure
exists in the secondary side of the steam generator.
Date of issuance: November 9, 1994
Effective date: November 9, 1994
Amendment Nos.: 151 and 133
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 28, 1994 (59
FR 49426) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 9, 1994. No significant
hazards consideration comments received: No.
Local Public Document Room location: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223
Florida Power and Light Company, Docket No. 50-335, St. Lucie
Plant, Unit No. 1, St. Lucie County, Florida
Date of application for amendment: March 19, 1993, augmented August
18, 1994.
Brief description of amendment: This amendment allows a reduction
in Reactor Coolant System design flowrate from the current value of
370,000 gpm to 355,000 gpm in Technical Specifications Figure 2.1-1 and
Tables 2.2-1 and 3.2-1.
Date of issuance: November 25, 1994
Effective date: November 25, 1994
Amendment No.: 130
Facility Operating License No. DPR-67: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 28, 1993 (58 FR
25855) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 25, 1994. No
significant hazards consideration comments received: No.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas
Date of amendment request: April 29, 1994, as supplemented by
letter dated September 8, 1994.
Brief description of amendments: The amendments revise the
technical specifications to permit revision of the maximum allowable
power range neutron flux high setpoint when one or more main steam
safety valves are inoperable. In addition, new algorithm used to
calculate the revised setpoint values is incorporated into the Bases
for the technical specifications.
Date of issuance: November 22, 1994
Effective date: To be implemented within 30 days of issuance
Amendment Nos.: Unit 1 - Amendment No. 66; Unit 2 - Amendment No.
55
Facility Operating License Nos. NPF-76 and NPF-80. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 8, 1994 (59 FR
29628) The additional information contained in the supplemental letter
dated September 8, 1994, was clarifying in nature and, thus, within the
scope of the initial notice and did not affect the staff's proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated November 22, 1994.No significant hazards consideration
comments received: No
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy
Center,Linn County, Iowa
Date of applications for amendment: June 4, 1993, as supplemented
February 4, 1994, and May 6, 1994.
Brief description of amendment: The amendment revised the Technical
Specifications (TS) by changing the requirements of the TS Section 3.6,
``Primary Systems Boundary,'' adding definitions into Section 1.0,
``Definitions,'' and revising Bases Section 3/4.6. These changes
improved clarity and provided consistency of the TS with the Standard
TS (NUREG-1202). Typographical and administrative corrections were also
made in Section 3.6.
Date of issuance: November 17, 1994
Effective date: November 17, 1994, and to be implemented within 120
days
Amendment No.: 203
Facility Operating License No. DPR-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 21, 1993 (58 FR
39052). The May 6, 1994, application, repeated a TS change included in
the June 4, 1993, application, and proposed changes to the TS Bases.
The information in the February 14, 1994, supplement, did not change
theinitial no significant hazards determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated November 17, 1994. No significant hazards consideration comments
received: No.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, SE., Cedar Rapids, Iowa 52401.
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center,
Linn County, Iowa
Date of application for amendment: July 12, 1994
Brief description of amendment: The proposed amendment changes the
requirement to perform the surveillance test for the channel functional
test Rod Block Monitor, Flow-biased Average Power Range Monitor and
Recirculation Flow instruments from within 24 hours prior to startup to
after the reactor is in the RUN mode, but prior to when each system is
assumed to function in the plant safety analysis.
Date of issuance: November 18, 1994
Effective date: November 18, 1994, to be implemented within 30 days
of issuance
Amendment No.: 204
Facility Operating License No. DPR-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 31, 1994 (59 FR
45025) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 18, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, SE., Cedar Rapids, Iowa 52401
South Carolina Electric & Gas Company, South Carolina Public
ServiceAuthority, Docket No. 50-395, Virgil C. Summer Nuclear
Station, Unit No. 1, Fairfield County, South Carolina
Date of application for amendment: October 29, 1993, as
supplemented March 11, 1994, May 18, 1994, September 20, 1994, and
October 20, 1994
Brief description of amendment: The proposed changes support the
installation of new steam generators at Summer Station. The changes
involve:
(1) alterations to the core operating limits
(2) changes to various reactor trip setpoints
(3) deletion of the negative flux rate trip
(4) removal of references to specific correlations used in the
departure from nucleate boiling (DNB) analyses
(5) changes to the steam/feedwater flow mismatch activation
specification
(6) changes to shutdown limits
(7) changes to instrument uncertainty allowances
(8) a change to the methodology for reactor coolant system (RCS)
flow determination
(9) modifications to DNB parameters
(10) a change to the engineered safety features actuation system
setpoints for steam generator water levels
(11) removal of the F* and L* criteria (12) addition of a
requirement for a first inservice inspection for the new steam
generators
Date of issuance: November 18, 1994
Effective date: November 18, 1994
Amendment No.: 119
Facility Operating License No. NPF-12. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 16, 1994 (59
FR 7968) The May 18, 1994, September 20, 1994, and October 20, 1994
submittals contained explanatory information and did not change finding
of nos significant hazards consideration as published in the FEDERAL
REGISTER. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 18, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180
South Carolina Electric & Gas Company, South Carolina Public
ServiceAuthority, Docket No. 50-395, Virgil C. Summer Nuclear
Station, Unit No. 1, Fairfield County, South Carolina
Date of application for amendment: July 20, 1994
Brief description of amendment: The amendment changes the Technical
Specifications (TS) to modify TS Table 2.2-1, Reactor Trip System
Instrumentation Setpoints, and Table 3.3-4, Engineered Safety Features
Actuation System Instrumentation Trip Setpoints and several associated
bases. The change would remove specific rack and sensor allowable drift
values by removing three columns from the tables.
Date of issuance: November 18, 1994
Effective date: November 18, 1994
Amendment No.: 120
Facility Operating License No. NPF-12. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: September 19, 1994 (59
FR 47181) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 18, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180
Southern California Edison Company, et al., Docket Nos. 50-361 and
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3,
San Diego County, California
Date of application for amendments: March 5, 1993, as supplemented
by letter dated September 22, 1994
Brief description of amendments: These amendments propose to revise
Technical Specification (TS) 3/4.7.1.1, ``Main Steam Safety Valves,''
and the associated Bases to (1) increase the as-found setpoint
tolerance of Table 3.7-1 for the Main Steam Safety Valves (MSSVs) from
+/- 1 percent to +2 percent and -3 percent; (2) add a footnote to Table
3.7-1 to indicate that the setpoint tolerance for the lowest set pair
of MSSVs will be +1 percent and -3 percent; (3) add a footnote to TS
3.7.1.1 and revise footnote 1 of Table 3.7-1 to clarify that the MSSVs
will be left at the lift setting according to Table 3.7-1 within a +/-1
percent tolerance following inservice testing; (4) add an ACTION
statement requiring the plant to be in HOT STANDBY within 6 hours and
in HOT SHUTDOWN within the following 12 hours for the case of less than
five MSSVs operable per operable steam generator; (5) require the plant
to be in ``HOT SHUTDOWN within the following 12 hours'' instead of
``COLD SHUTDOWN within the following 30 hours'' per the existing ACTION
statement; (6) revise the title of column 1 of Table 3.7-2 to read
``Number of Operable Safety Valves per Operable Steam Generator,''
instead of ``Maximum Number of Inoperable Safety Valves on Any
Operating Steam Generator, for better readability; and (7) delete the
ORIFICE SIZE column of Table 3.7-1.
Date of issuance: November 23, 1994
Effective date: As of the date of its issuance and must be fully
implemented no later than 30 days from the date of issuance.
Amendment Nos.: 114 and 103
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 23, 1993 (58 FR
34093) The additional information contained in the supplemental letter
dated September 22, 1994, served to clarify the amendments, was within
the scope of the initial notice, and did not affect the Commission's
proposed no significant hazards consideration determination.The
Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated November 23, 1994.No significant hazards
consideration comments received: No.
Local Public Document Room location: Main Library, University of
California, P. O. Box 19557, Irvine, California 92713
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: September 9, 1994 (TS 94-08)
Brief description of amendments: The amendments add the main steam
valve vaults to the exclusion areas where containment penetration
integrity is not required to be verified once every 31 days for
penetrations that are secured in the closed position.
Date of issuance: November 22, 1994
Effective date: November 22, 1994
Amendment Nos.: 191 and 183
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: October 12, 1994 (59 FR
51630) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 22, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia.
Date of application for amendments: July 14, 1994
Brief description of amendments: These amendments modify the
current Technical Specifications having cycle-specific parameter limits
in the Core Operating Limits Report.
Date of issuance: November 15, 1994
Effective date: November 15, 1994Amendment Nos. 194 and 194
Facility Operating Licen se Nos. DPR-32 and DPR-37: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 12, 1994 (59 FR
51630) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 15, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
NuclearPower Plant, Kewaunee County, Wisconsin
Date of application for amendment: May 17, 1994
Brief description of amendment: The amendment revises the TS by
separating the specification for the Internal Containment Spray (ICS)
and the Spray Additive Systems into two distinct specifications. The
amendment also removes the requirement that for a spray train to be
operable, a spray pump suction flow path from the additive tank is
needed. In addition, the allowable out-of-service time for the Spray
Additive System is increased from 48 hours to 72 hours.
Date of issuance: November 18, 1994
Effective date: November 18, 1994
Amendment No.: 113
Facility Operating License No. DPR-43. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 20, 1994 (59 FR
37090) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 18, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: University of Wisconsin
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin
54301.
Dated at Rockville, Maryland, this 30th day of November 1994.
For the Nuclear Regulatory Commission
Jack W. Roe,
Director, Division of Reactor Projects - III/IV,Office of Nuclear
Reactor Regulation
[FR Doc. 94-29925 Filed 12-6-94; 8:45 am]
BILLING CODE 7590-01-F
